ML20081J197

Forwards 10CFR50.46 Annual ECCS Evaluation Model Changes for 1994

ML20081J197
Person / Time
Site:	Farley
Issue date:	03/20/1995
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9503270145
Download: ML20081J197 (14)

Text

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~ Southem Nucioar Opecting Company

'8 Fost Offc) Box 1295 Bemingha"t, Alzb;.ma 35201

' Telephone (205) 868-5131 Southem Nudear Operating Company o m u ey Nr*ehr*h*c"t thrch 20,1995 the Southem electnc system 10 CFR 50.46 Docket Nos.: 50-348 50-364 U. S. Nuclear Regulatory Commission.

ATTN: Document Control Desk Washington, D. C. 20555 Joseph M. Farley Nuclear Plant - Units 1 and 2 10 CFR 50.46 Annual ECCS Evalyation Model Changes Report for 1994 Gentlemen:

Provisions in 10 CFR 50.46 require applicants and holders of operating licenses or construction permits to annually notify the Nuclear Regulatory Commission (NRC) of changes and errors in the Emergency Core Cooling System (ECCS) Evaluation Models. In compliance with this requirement, enclosed is the Southern Nuclear Operating Company's report for Joseph M. Farley Nuclear Plant Units I and 2 for the calendar year 1994.

The annual report provides information regarding the effects of the ECCS Evaluation Model modifications on the peak cladding temperature (PCT) results since the 1993 annual report. Also, the attached annual report provides a summary of the plant changes performed under the provisions of 10 CFR 50.59 that also affect the PCT results. The report is in accordance with the Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting (WCAP-13451).

It has been determined that compliance with the requirements of 10 CFR 50.46 continues to be maintained when the effects of plant design changes are combined with the effects of the ECCS Evaluation Model changes and errors applicable to Farley Units 1 and 2.

If there are any questions, please advise.

Respectfully submitted,

() 9 hi.

D. N. Morey REM / cit:ccesmods. doc Attachment cc: Mr. S. D. Ebneter Mr. B. L. Siegel Mr.T. M. 'Ross

%3 9503270145 950320 PDR ADOCK 05000348 l

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j ATTACHMENT JOSEPH M. FARLEY NUCLEAR PLANT 10 CFR 50.46 ECCS EVALUATION MODEL 1994 ANNUAL REPORT L

BACKGROUND Provisions in 10 CFR 50.46 require applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory Commission (NRC) of errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models on an annual basis. 10 CFR 50.46 requires that significant errors or changes in the ECCS Evaluation Model be reported to the NRC within 30 days with a proposed schedule for providing a re-analysis or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements. 10 CFR 50.46 defines a significant error or change as one which results in a calculated fuel peak cladding temperature (PCT) different by more than 50*F from the temperature calculated for the limiting transient using the last acceptable model, or as a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50*F.

In Reference 1, information was submitted to the NRC regarding modifications to the Westinghouse large-break and small-break Loss-of-Coolant Accident (LOCA) ECCS Evaluation Models as applicable to the Farley Nuclear Plant (FNP) analyses for the calendar year 1993. In Reference 2, a significant error report on the small-break LOCA Evaluation Model results for Farley Units 1 and 2 was submitted to the NRC.

The following presents an assessment of the effects of modifications to the Westinghouse ECCS Evaluation Models on the Farley LOCA analysis results since the 1993 annual report (Reference

1) for the calendar year 1994. The 1994 annual report also reflects the recent re-analysis of the Unit I large-break LOCA performed in 1994 to support the Unit 1 1.70 FAH license amendment. This annual report has been prepared in accordance with the Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting (WCAP-13451, Reference 3).

The results presented in the annual report as an analysis-of-record for the large-break LOCA and small-break LOCA PCTs reflect the use of VANTAGE-5 fuel in both units (Reference 4).

IL LARGE-BREAK LOCA Table I shows the large-break LOCA PCT rackups for both Unit I and Unit 2.

ILA LARGE-BREAK LOCA ANALYSIS-OF-RECORD The large-break LOCA analyses for Farley Units 1 and 2 were examined to assess the effects of l

the changes and errors in the Westinghouse large-break LOCA ECCS Evaluation Model on PCT results.

The large-break LOCA analysis-of-record results for Farley Units 1 and 2 were calculated using the 1981 version of the Westinghouse large-break LOCA ECCS Evaluation Model 1

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A'ITACHMENT j

Page 2 incorporating the B ASH analysis tecimology (Reference 5). The large-break LOCA analysis for Unit I was revised in 1994 to support a 1.70 FAH licensing amendment for Unit 1 (Reference 6).

The Unit I and Unit 2 analyses assumed the following information important to the large-break LOCA analyses:

Unit 1 Unit 2 Core Power = 1.02 x 2652 MWT Core Power = 1.02 x 2652 MWT 17x17 VANTAGE-5 Fuel Assembly 17x17 VANTAGE-5 Fuel Assembly Fo = 2.45 for VANTAGE-5 Fuel Fo = 2.45 for VANTAGE-5 Fuel Fo = 2.32 for LOPAR Fuel Fo = 2.32 for LOPAR Fuel FAf! = 1.70 for VANTAGE-5 Fuel FAH = 1.65 for VANTAGE-5 Fuel FAH = 1.55 for LOPAR Fuel FAH = 1.55 for LOPAR Fuel SGTP* = 20%

SGTP* = 20%

Upflow Configuration Downflow Configuration

• SGTP = Steam generator tube plugging limit assumed in the LOCA analysis For Farley Unha 1 and 2, the limiting size break analysis-of-record is a double-ended guillotine mpture of the cold leg piping with a discharge coefficient of Co = 0.4. The limiting PCTs determined for the Unit I and Unit 2 large-break are 19637 and 2141*F, respectively. Both the Unit I and Unit 2 analysis-of-record limiting PCT values include 37 for containment mini-purge automatic isolation, 87 for increased Tavg temperature uncertainty, and 6*F for combined safe shutdown earthquake (SSE) and LOCA events. Also included in the limiting PCT values for both units is the addition of a 507 transition penalty due to the mixed core conditions during the transition to VANTAGE-5 fuel. However, the above penalties are listed in Table 1 according to the format of WCAP-13451 (Reference 3) and are listed separately because they are not explicitly modeled in the ECCS analysis.

l ILB 199410 CFR 50,46 LOCA MODEL ASSESSMENTS The following changes and errors in the Westinghouse ECCS Evaluation Models would affect the BASH Evaluation Model large-break LOCA analysis-of-record results.

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ATTACHMENT Page 3 II.B.1 Prior Reported Assenments The prior large-break LOCA PCT assessments given in Table 1 were submitted to the NRC in March 1994 as part of the 1993 Annual Report (Reference 1). It is noted in Table 1 that the structural metal heating correction was explicitly accounted for in the recent re-analysis of the large-break LOCA for Unit 1 (Reference 6).

II.B.2 Cp_de Stream Improvement (0*F)

Revisions were made to the procedures used to interface the various codes that comprise the entire execution stream for performing a large-break LOCA analysis with the BASH Evaluation Model.

The previous use of the coupled WREFLOOD/ COCO code for calculating containment pressure response, which was then transferred as a boundary condition to the BASH code, has been replaced with direct coupling of the BASH and COCO codes, such that the same code used to calculate the RCS conditions during reflood also supplies the boundary conditions for the containment pressure calculation. In conjunction with this, the portion of the WREFLOOD code which calculated the refill phase of the transient has been reprogrammed into a separate, but identical code called REFILL, which is also coupled with COCO.

This methodology revision was made only as a process improvement for conducting analyses and involved no changes to the approved physical models nor basic solution techniques governing the solutions provided by the individual computer codes. The NRC has been advised by Westinghouse of the implementation of this methodology on a forward-fit basis.

Due to small perturbations in the boundary conditions resulting from this revised methodology for interfacing the codes, small differences in predicted results were observed. The effects were minor, with no observed bias. Since this methodology is a process improvement which is to be implemented on a forward-fit basis, there are no effects on existing licensing analyses, and any small effects on results will be implicitly accounted for in future analyses. It is noted that this improved methodology was used in the recent Unit I re-analysis of the large-break LOCA. The impact on the Unit 2 PCT results is 0*F.

II.B.3 Loop / Core Interface Corrections in B ASH (0*F)

Corrections were made to the logic for interfacing the loop model and BART code model. One correction prevents the possibility of an occasional inconsistency in how the core time step was limited by the loop time step. Another corrects the fluid density used in the interface calculation when the inlet tiow rate is negative.

Results from sensitivity studies for the corrections demonstrated negligible perturbations in the trends of the system parameters with a very minor net effect on peak clad temperature predictions relative to the results from the previous version. Since this is an extremely small effect, with no 1

.o ATTACHMENT Page 4 apparent bias, the net effect on existing analyses is estimated to be 07 for margin tracking purposes. The change has been implemented on a forward-fit basis only and will be incorporated implicitly in any future analyses.

II.B.4 Pellet Power Radial Flux Depression Error (OT)

A coding error (an incorrect sign) was discc,vered and corrected in a subroutine that calculates radial distribution power factors in the fuel pellet for the LOCBART code.

Sensitivity studies found the error correction to result in less than a 10.lT effect on predicted peak clad temperature. The net effect on existing analyses is therefore 07 for margin tracking purposes and will be implicitly included in future recalculations.

H.B.5 trpprovements to Flooding Rate Smoothing (OT)

Part of the approved methodology for performing the large-break LOCA analyses with the BASH Evaluation Model is the requirement that the core inlet flooding rate calculated by the BASH code be linearized in a piece-wise manner to remove oscillations prior to use in the hot channel fuel rod calculation. This operation is termed " smoothing," and guidelines are provided to the analysts describing how to linearize the curve by observing inflections in slee overall flooding rate. To facilitate consistency in performing this operation, the logic has been coded into a program named SMUUTH. A new version of the SMUUTH program has been implemented which incorporates improved logic for determining the inflection points gained through experience in utilizing the program for a broad range of plant transients.

There are no changes to the approved evaluation model methooology from this revision. The SMUUTH program merely represents a convenient way of automating the approved methodology and does not explicitly introduce any effects on the results. T'ais revision is being reported only as a change to the code stream used for standard analyses. There are OT effects on the predicted results from using the new program version. The new method was used in the recent Unit I re-analysis of the large-break LOCA (Reference 6).

II.B.6 Accumulator Water Temperature (OT)

The choice of accumulator water temperature can affect the calculated Peak Cladding Temperature (PCT) associated with the ',arge-break LOCA analyses. Early Westinghouse Evaluation Models had assumed a generic voue of 907 for the accumulator water temperature based on a conservatively low valae of containment air temperature at 100% power in fulfillment of the Appendix K requirements associated with the calculation of a low containment back pressure.

These containment initial temperature and pressure assumptions in a plant's large-break LOCA analysis have been consistently reported to the NRC in the FSAR. The NRC had previously

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.s ATTACHMENT Page 5 reviewed and approved this aspect of the large-break LOCA Evaluation Model via plant-specific safety evaluation reports. Using these assumptions, and with the early Westinghouse models,90T was conservative with respect to the overall effect on the large-break LOCA PCT.

Newer evaluation models have demonstrated that a higher containment air temperature, coupled with higher accumulator water temperatures, may result in an even more conservative calculation for PCT, even if containment pressure is slightly higher than calculated with the 90*F assumption.

Sensitivity studies performed with the newer evaluation models (identified below) have shown a small sensitivity to accumulator water temperature. The effect on PCT was a 1.3T change in PCT for a IT change in accumulator water temperature when the accumulator water temperature varies over a range from 90*F to 120*F. Application of this sensitivity over its applicable range results in a PCT effect which is below the 10 CFR 50.46 threshold for determination of a significant change (i.e., < 507). It is therefore Southern Nuclear's position that immediate implementation of this new methodology is not required. As such, application of the new plant-specific methodology and associated change in analysis assumptions can be fonvard-fit to new large-break LOCA analyses.

In support of future analyses, a set of criteria has been developed for selection of the accumulator water temperature for use in large-break LOCA analyses which use either the 1981 Evaluation Model with BART or the 1981 Evaluation Model with BASH. These data will be utilized for future large-break LOCA analyses.

H.C 10 CFR 50.59 SAFETY EVALUATIONS FOR NON-MODEL IMPACTS No 10 CFR 50.59 safety evaluations for non-model impacts have been assessed against the reference analysis results to date.

H.D TOTAL RESULTANT LARGE-BREAK LOCA PCT As discussed above, the changes and errors to the Westinghouse large-break LOCA ECCS Evaluation Model could affect the large-break LOCA analysis results by altering the PCT. As shown in Table 1, the large-break LOCA analysis PCT results for both units are below the 10 CFR 50.46 limit of 22000F.

II.E LARGE-BREAK LOCA CONCLUSIONS An evaluation of the effects of changes and errors in the Westinghouse large-break BASH ECCS Evaluation Model was performed on the large-break LOCA applicable to the Farley reference analysis. When the efTects of the large-break ECCS Evaluation Model changes and errors were combined with those of plant changes and the large-break LOCA analysis-of-record results, it was determined that Farley Units I and 2 were in compliance with the requirements of 10 CFR 50.46.

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ATTACHMENT Page 6 III.

SMALI BREAK LOCA Table 2 shows the small-break LOCA PCT rackups for both Unit I and Unit 2.

III.A SMAL1rBREAK LOCA ANALYSIS-OF-RECORD The small-break LOCA analyses for Farley Units 1 and 2 were also examined to assess the effects of the changes and errors to the Westinghouse small-break LOCA ECCS Evaluation Models on PCT results. The small-bredc LOCA ECCS analysis results were calculated using the NOTRUMP small-break LOCA ECCS Evaluation Model (Reference 7).

The Unit I and Unit 2 analyses assumed the following information important to the small-break LOCA analyses:

Unit 1 Unit 2 Core Power = 1.02 X 2775 MWT Core Power = 1.02 x 2775 MWT 17x17 VANTAGE-5 Fuel Assembly 17x17 VANTAGE-5 Fuel Assembly FQ = 2.50 FQ = 2.50 FAH = 1.70 FAH = 1.70 Upflow Configuration Downflow Configuration For Farley Units I and 2, the limiting size break analysis-of-record for the VANTAGE-5 fuel analysis is a 3-inch diameter break in the cold leg. The limiting PCTs determined for the Unit I and Unit 217x17 VANTAGE-5 small-break are 1805'F and 1783*F, respectively. Both the Unit I and Unit 2 analysis-of-record limiting PCT values include a 20*F penalty due to the increased Tavg temperature uncertainty. However, the above penalties are listed in Table 2 according to the format of WCAP-13451 (Reference 3) and are listed separately because they are not explicitly modeled in the ECCS analysis.

III.B 199410 CFR 50.46 LOCA MODEL ASSESSMENTS The following changes and errors in the Westinghouse ECCS Evaluation Models would affect the NOTRUMP small-break LOCA analysis results obtained for the Farley VANTAGE-5 fuel analysis. Information provided herein in items III.B.2, III.B.3, III.B.4, and III.B.5 were previously provided to the NRC in November 1994 (Reference 2). However, they are repeated herein for completeness.

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A~ITACIIMENT Page 7 III.B.1 Prior Reconed Assessments The prior small-break LOCA PCT assessments shown in Table 2 were submitted to the NRC in November of 1994 (Reference 2). The information concerning the prior assessments in Reference 2 was based on information contained in the 1993 annual report (Reference 1).

III.B.2 Boiling Ilcat Transfer Correlation Errors (-6%

This closely related set of errors deals with how the mixture velocity is defined for use in various boiling heat transfer regime correlations. The previous definition for mixture velocity did not properly account for drift and slip effects that were calculated using NOTRUMP. Particularly affected were the NOTRUMP calculations of heat transfer coefficient when using the Westinghouse Transition Boiling Correlation and the Dougall-Rohsenow Saturated Film Boiling Correlation. In addition, a typographical error was also corrected in the Westinghouse Transition Boiling Correlation. These errors resulted in a 6*F benefit on PCT for both Unit I and Unit 2.

See Note A below for impact on the Burst and Blockage / Time in Life penalty.

III.B.3 Main Steam Isolation Logic Error (18%

This error consists of two ponions: 1) a possible plant-specific effect which was the result of incorrect logic which caused the main steam line isolation to occur on the same signal as main feedwater isolation on S-signal, which is inconsistent with the standard conservative assumption of steam line isolation on Loss of Offsite Power coincident with the earlier reactor trip signal, and 2) a generic effect which was applicable to all previous analyses and was the result of incorrect logic which always led to isolation functions occurring at a slightly later time than when the appropriate signal was actually generated. The plant-specific efTect is not applicable to Farley; however, the generic efTect resulted in an 18*F penalty on PCT for both Unit I and Unit 2.

See Note A below for impact on the Burst and Blockagefrime in Life penalty.

III.B.4 Core Node Zire Oxide Initialization Error (0*F.)

The NOTRUMP code models two regions for each core node analogous to the two regions (mixture and vapor) in adjoining fluid nodes. During the course of a transient, NOTRUMP tracks region-specific quantities for each core node. Erroneous logic caused incorrect initialization of the region-specific fuel cladding zirc oxide thickness at times prior to the actual creation of the relevant region during the core boil-off transient. However, this error had a 0*F cffect on PCT.

l See Note A below for impact on the Burst and Blockage / Time in Life penalty.

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- ATTACHMENT Page 8, Note Ai Associated with the above NOTRUMP errors is a 107 increase in the current Burst and Blockage / Time in Life penalty for Farley Unit I and a 87 increase in the current Burst and Blockage / Time in Life penalty for Farley Unit 2. Note that the Burst and Blockagerrime in Life penalty is a function of PCT.

III.B.5 Axial NWh=4n. RIP Model Revision. SBLOCTA Error Corrections Analysis (14T for Unit 1 and -497 for Unit 2)

The standard rod model developed by Westinghouse in the 1970's whicit was used to perform the SBLOCTA calculations has 19 axial nodes with a finer distribution in the top elevations.

However, sensitivity studies performed by Westinghouse to justify the number and distribution of these nodes could not be documented. Therefere, a series of calculations were performed using increasingly finer axial nodalizations than prescribed for the 19-node model and indicated that the standard SBLOCTA 19-node model was not conservative. Nearly all cases demonstrated a significantly non-conservative behavior with respect to PCT. The penalty is attributed to a net increase in single phase steam enthalpy rise as these nodes uncover sooner and heat up more than the coarser nodes partially covered by the mixture level. It was concluded that a revised model that included a much finer axial nodalization could potentially lead to less favorable results than those predicted in the current analyses. Thus, a revised standard for rod nodalization has been established that ensures an adequate solution to the hot channel calculation by specifying a fine nodalization of 0.25 ft. nodes for all elevations that are predicted to uncover during the transient.

As a separate, but related issue, Westinghouse implemented a revised model for calculating transient fuel rod internal pressure in the SBLOCTA code. Fuel rod pressure is a governing factor in defining the clad creep, burst and blockage behavior for SBLOCA transients. The NRC has been informed of this modeling change, along with the fact that Westinghouse has validated and instituted the model as a methodology improvement to the SBLOCA model for standard implementation on a forward-fit basis.

Since the improved axial nodalization methodology and revised fuel rod internal pressure model can have significant synergistic effects on the predicted PCT, the SBLOCTA calculation from the limiting SBLOCA transient has been rerun with the revised code and methodology in order to obtain an accurate estimation of the net effect of these changes on the analysis-of-record.

Several recent code revisions and error corrections oflesser magnitude were also incorporated into the code version used by Westinghouse to conduct this calculation. Normally, these items would have been reported in the 10 CFR 50.46 year-end reporting summary along with estimates of the effects. As a consequence of using the revised code to obtain the results for this l

evaluation, these items were implicitly addressed by Westinghouse in the results provided. Since all of the issues relate to portions of the SBLOCTA code and its associated methodology, they are reported as a single closely-related group of changes under " Axial Nodalization, RIP Model Revision, and SBLOCTA Error Corrections." The re-analysis of the limiting break size using the improved SBLOCTA code and methodology (with 34 axial nodes) resulted in a 14T penalty for Farley Unit I and a 497 benefit for Farley Unit 2.

.. o NITACHMENT Page 9,

Note B: As discussed in Note A above, a Burst and Blockage /fime in Life penalty / benefit has also been appropriately assessed (ST penalty for Farley Unit I and a 28'F benefit for Farley Unit 2) to the SBLOCTA nodalization error.

III.B.6 Pressure Search Convergence Criteria in NOTRUME.(0*E)

The convergence criteria used during the pressure search in NOTRUMP have been found to be not adequately restrictive to ensure a sufficiently accurate value for Fluid Node pressure when conditions approach the boundary between subcooled and saturated in some cases. The resulting effects on predicted pressure were more pronounced at pressures below those normally

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seen during standard Evaluation Model calculations. The previously hardwired convergence criteria values have been made by user input, appropriate values have been determined, and these will be implemented in all future analyses. However, this change had a 0*F effect on PCT.

III.B.7 Friction Value Input Correction in NOTRUMP (0*F)

The SPADES code is used to generate input decks for the small-break analysis code, NOTRUMP. An error was found in the code which involved the values assigned to some of the friction factor input. The erroneous values had no impact on transient calculations and were corrected in order to maintain the consistency of the SPADES code with the relevant documentation. Thus, the effect of the above correction had a 0*F effect on PCT.

l III.B.8 Aulomatic Containment Sprav Actuation in NOTRUMP (0*F) l l

Automatic containment spray actuation during a small-break LOCA had not previously been l

addressed in the Westinghouse small-break LOCA evaluation model. The containment pressure l

transient is not modeled because the small-break PCT is net directly sensitive to this effect.

l While investigating this issue, however, Westinghouse concluded that containment spray l

actuation early in the small-break transient is possible for a variety of containment types.

Containment spray actuation could result in draindown of the RWST prior to conclusion of the small-break transient. Switching to cold leg recirculation during the transient may reduce or briefly interrupt the modeled ECCS injection flow in some plants and elevate the enthalpy of the ECCS injection water. Furthermore, an alternate single failure scenario could result in earlier l

draindown for the RWST and subsequent switchover to cold leg recirculation. This issue had a 0*F cffect on PCT at Farley since automatic spray actuation does not occur on a small-break 1

LOCA.

Ill.B.9 Safety Iniection in the Broken Loop (0*F)

WCAP-10054-P, Addendum 2," Addendum to the Westinghouse Small-Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," August 1994, was recently submitted by Westinghouse to the NRC.

ATTACHMENT Page 10 The above topical repon presents a change to the Westinghouse small-break LOCA methodology dealing with ECCS flows in the broken loop. It also presents a revised condensation model that will be used on the safety injectionjet in future analyses. References I and 2 provide additional discussions concerning the above two issues regarding SI in the broken loop and the COSI condensation model.

This change is being implemented on a forward-fit basis prior to formal approval in accordance with Section 4.1.3 of WCAP-13451 (Reference 3). The change has been shown to typically produce PCT benefits in studies presented in the above topical report. Since it is being implemented on a forward-fit basis, a net PCT impact of 0*F is being assessed against existing analyses.

III.C 10 CFR 50.59 SAFETY EVALUATIONS FOR NON-MODEL IMPACTS No 10 CFR 50.59 safety evaluations for non-model impacts have been assessed against the reference VANTAGE-5 LOCA analysis results to date. It should be noted that the effects of all of the applicable previous evaluations for both Farley Units 1 and 2 were incorporated into the VANTAGE-5 analysis.

II!.D TOTAL RESULTANT SMALI BREAK LOCA PCT As discussed above, the changes and errors in the Westinghouse small-break LOCA ECCS Evaluation Model could affect the small-break LOCA analysis results by altering the PCT as shown in Table 2.

III.E SMALI BREAK LOCA CONCLUSIONS An evaluation of the effects of changes and errors to the Westinghouse ECCS Evaluation Model was performed for the small-break LOCA analysis results. When the effects of the small-break ECCS Evaluation Model changes and errors were combined with those of plant changes and the small-break LOCA analysis-of-record results, it was determined that compliance with the requirements of 10 CFR 50.46 would be maintained for both Units 1 and 2.

IV.

REFERENCES

1. Letter from D. N. Morey to USNRC, " Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1993," March 18,1994.

2. Letter from D. N. Morey to USNRC, " Joseph M. Farley Nuclear Plant Peak Clad Temperature (PCT) Calculation," November 30,1994.

3. WCAP-13451, " Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," dated October 1992.

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..a ATTACHMENT Page ll, i

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4. NRC Safety Evaluation Report, " Issuance of Amendmem No. 92 to Facility Operating License No. NPF-2 and Amendment No. 85 to Facility Operating License No. NPF-8 Regarding the Use of VANTAGE-5 Fuel in Both Units and Allowing Removal and Replacement of the Resistance Temperature Detector Bypass Manifold System in Unit 2 -

Joseph M. Farley Nuclear Plant, Units I and 2 (TAC Nos. M81025 and M81026)," March 11,1992.

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- 5. "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code,"

s WCAP-10266-P-A, Rev. 2 (Proprietary), Young, M. Y., et. al, March 1987.

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6. NRC Safety Evaluation Report, " Issuance of Amendment No.109 to Facility Operating License No. NPF-2 Regarding Nuclear Enthalpy Rise Hot Channel Factor - Joseph M.

i Farley Nuclear Plant, Unit 1 (TAC No. M89665)," July 22,1994.

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7. " Westinghouse Small-break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary), Lee, N., et. al, August 1985.

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~..,,, a TABLE 1

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JOSEPH M. FARLEY NUCLEAR PLANT

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TOTAL RESULTANT LARGE-BREAK LOCA PCT (019 i

t A. ANALYSIS-OF-RECORD (VANTAGE-5)

Unit L *F Unit 2.7 l

1. ECCS Analysis 1896*

2074*

2. Containment Mini-Purge Auto Isolation 3

3 l

3. Tavg Temperature Uncertainty 8

8

4. Combined SSE and LOCA Events 6

6

5. Transition Core Penalty 50 50 Total Analysis-of-Record PCT =

1963*

2141*

i E. 199310 CFR 50.46 MODEL ASSESSMENTS

1. PriorReported Assessments 6**

- 31 *

• i C.10 CFR 50.59 PLANT MODIFICATIONS i

None 0

0 i

D. TOTAL RESULTANT LARGE-BREAK LOCA PCT 1957 2110 i

i t

The PCT values are rounded up to the next highest integer number to avoid reporting in decimal points. The Unit I values correspond to an FAH increase to 1.70 (Reference 7).

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The Structural Metal Heat Modeling Correction (-250F) and the LUCIFER error l

correction (-6'F) were submitted to the NRC in March 1994 as part of the 1993 Annual Report (Reference 1). However, in the recent re-analysis of Unit 1 to increase the FAH from 1.65 to 1.70, the structural metal heat modeling was explicitly accounted for in the large-break LOCA for Unit 1 (Reference 7).

..;s a.s TABLE 2 -

i JOSEPH M. FARLEY NUCLEAR PLANT TOTAL RESULTANT SMALL-BREAK LOCA PCT (DF) i A.

ANALYSIS-OF-RECORD (VANTAGE-5)

Unit 1.T-U nit 2. T

1. ECCS Analysis 1785*

1763*

l

2. Tavg Temperature Uncertainty

__20

_2Q Total Analysis-of-Record PCT =

1805 1783 B.

199310 CFR 50.46 MODEL ASSESSMENTS

1. ~ Prior Reported Assessments 130*

113*

2. Boiling Heat Transfer Correlation Error 6**

6"

3. Steam Line Isolation Logic Error 18 "

18**

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4. Axial Nodalization, RIP Model Revision, SBLOCA Error Correction Analysis

• 14 "

- 49*

• l S. Change in Burst and Blockage /fime in Life

• 15 "

- 20" C.

10 CFR 50.59 PLANT MODIFICATIONS None 0

0 i

D.

TOTAL RESULTANT SMALL-BREAK LOCA PCT 1976 1839 i

4 i

Reported to the NRC under 10 CFR 50.46 in Reference 1.

i Reported to the NRC under significant reporting requirements of 10 CFR 50.46 in November of 1994 (Reference 2). These are repeated here for completeness.

The re-analysis of SBLOCTA also corrected the Average Rod Burst Strain (ST for Unit 1 and 4T for Unit 2) and the Fuel Rod Burst Strain Limit (-57 for Unit I and -4*F for Unit

2) PLOCTA corrections reported in Reference 1. The net effect for these two corrections is 07 for both units.

For Burst and Blockage /fime in Life, penalties of 52*F for Unit I and 35T for Unit 2 were included in B.1 above as previously reported to the NRC in References 1 and 2.

Item B.5 reflects changes to the reported values for this issue since the Burst and Blockage /fime in Life penalty is a function of PCT as reported to the NRC (Reference 2) in November 1994. (Repeated here for completeness.)

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